ML23059A457

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Presentation Slides - Periodic Advanced Reactor Stakeholder Meeting 03022023
ML23059A457
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Issue date: 03/01/2023
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Advanced Reactor Stakeholder Public Meeting March 2, 2023 Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 417 405 578#

Time Agenda Speaker 10:00 am - 10:15 am Opening Remarks / Adv. Rx Integrated Schedule / Update on SCALE/MELCOR NRC Advanced Reactor Source Term Demonstration Project 10:15 am - 10:55 am Advanced Reactor Construction Oversight Program (ARCOP) NRC 10:55 am - 11:30 am Advance Contracting Requirement Under Section 302(b) of the Nuclear Waste NRC / DOE Policy Act 11:30 pm - 12:00 pm Micro-Reactor Deployment Policy Topics NRC 12:00 pm - 1:00 pm Lunch Break All 1:00 pm - 1:30 pm Transportation and Storage for Advanced Reactor Fuel and Transportable NRC Micro-Reactors 1:30 pm - 1:45 pm Guidance for Reviewing a Non-Power Liquid Fueled Molten Salt Reactor License NRC Application 1:45 pm - 2:00 pm Pre-Application Engagement on Materials Qualification Issues for Advanced NRC Reactor Licensing 2:00 pm - 2:30 pm Advanced Reactor Materials Interim Staff Guidance NRC 2

Time Agenda (continued) Speaker 2:30 pm - 2:45 pm Break All 2:45 pm - 3:30 pm Status of Two Draft Regulatory Guides on RIPB Seismic Design and Seismic NRC Isolation for Commercial Nuclear Powerplants 3:30 pm - 3:35 pm Future Meeting Planning and Concluding Remarks NRC 3

Advanced Reactor Integrated Schedule of Activities The updated Advanced Reactor Integrated Schedule is publicly available on NRC Advanced Reactors website at:

https://www.nrc.gov/reactors/new-reactors/advanced/integrated-review-schedule.html 4

Advanced Reactor Integrated Schedule of Activities https://www.nrc.gov/reactors/new-reactors/advanced/integrated-review-schedule.html 5

Update on SCALE/MELCOR Advanced Reactor Source Term Demonstration Project

  • Developed new SCALE and MELCOR modeling capabilities for five non-light water reactor designs (2021-2022)
  • Held workshops that included sample accident simulations
  • Workshop documentation is now available on NRCs advanced reactor source term website
  • Held workshop on applying SCALE and MELCOR to the TRISO fuel cycle (February 28, 2023)
  • Will develop and demonstrate targeted model improvements (2023) 6

SCALE/MELCOR non-LWR source term demonstration project

  • Heat-pipe reactor workshop on June 29, 2021
  • Slides
  • Video Recording June 29, 2021
  • SCALE report
  • MELCOR report
  • High-temperature gas-cooled reactor workshop on July 20, 2021
  • Slides
  • Video Recording July 20, 2021
  • SCALE report
  • MELCOR report
  • Fluoride-salt-cooled high-temperature reactor workshop on September 14, 2021
  • Slides September 14, 2021
  • Video Recording
  • SCALE report
  • MELCOR report
  • Molten-salt-fueled reactor workshop on September 13, 2022
  • Slides September 13, 2022
  • Video Recording
  • SCALE report
  • Sodium-cooled fast reactor workshop on September 20, 2022
  • Slides September 20, 2022
  • Video Recording 7

Advanced Reactor Construction Oversight Program Division of Advanced Reactors and Non-Power Production and Utilization Facilities

To develop the best oversight program possible that ensures safety and security, considers the diversity of technology and its risk profile, adapts to facility-specific insights, and leverages our collective experience, while remaining adaptable to respond to future opportunities and challenges.

- ARCOP Challenge

Objective W H AT: Provide reasonable assurance that advanced reactor plants are built and will operate in accordance with their licenses and applicable laws and regulations, thus adequately protecting the public and environment H OW : Leverage an oversight program that is comprehensive, scalable, innovative, risk-informed, performance based, and technology-inclusive

What is New in ARCOP?

  • Project-specific inspection scope
  • Scalable inspection scope commensurate with performance
  • Focus on QA performance in different construction areas
  • Scalable inspection footprint including role of construction resident inspector
  • Streamlined significant determination process commensurate with facility risk
  • Performance assessment includes short term assessment for timely reaction to emergent issues
  • Explore the use of 3rd party performance monitoring data

Key Considerations in ARCOP Development Reactor Technology/ Manufacturing/

Attributes Construction Techniques Licensing Pathway o Different coolants, fuel, materials & o Greater use of factory fabrication o Parts 50, 52 &53 design codes/standards o Different information may be available during o Co-location with fuel facilities construction o Wide range of sizes o Will ensure consistency of oversight o Enhanced safety margin/risk profile 12

Building on Oversight Experience NUREG-1055 AP1000 SHINE FUEL FACILITIES DOE Greater focus on Enable flexibility, Value of hybrid Greater focus on Insights from quality assurance integration, and flexible design control, advanced reactor scalability & inspection scope procedures & construction hybrid capabilities procurement 13

Performance Monitoring Enhancements Scope Hybrid Inspection Technology/facility-specific Enables flexibility without sacrificing quality Anchored to fundamental safety functions Optimizes inspection conduct Considers risk-insights for reactor & SSCs Leverages technology QA + direct SSCs inspections Operational readiness & security Schedule Resources Project-specific Flexible, matches construction/manufacturing pace Commensurate with facility complexity and size Supports inspection at different locations Enables frequent performance assessment & scope Reflective of FOAK vs DOE proven technology adjustment 3rd Party Oversight RTR Inspection Insights Gain additional data Apply ARCOP to near-term RTR construction and Reduce redundancy refine based on experience Leverage international inspections Credit Authorized Inspection Agencies 14

Construction Inspection Matrix (example)

Fundamental Safety Functions Procurement of ASME Manufacturing of Reactor Construction of Steel &

Qualified Piping Vessel & Internals Concrete Buildings Reactivity Control SSC1, SSC2 inspection family SSC3, SSC4 N/A Decay Heat Removal SSC1, SSC5 SSC3, SSC4, SSC6 SSC7, SSC8 Radioactive Material Retention SSC1, SSC2 SSC9, SSC10 SSC11, SSC12 15

ARCOP Vertical Slice Inspection Risk-Significant SSC

01. In-depth inspection of QAP attributes Design Control associated with sampled SSCs. Shipping Procedures Quality Assurance Attributes
02. Optimizes inspection strategies used for vendors and SHINE Procurement Material Spec
03. Results inform assessment of construction area adequacy Testing Repeat for CAP Inspect SSC other SSCs in Choose risk Adjust AND same Documentation significant baseline inspect construction SSC for inspections applicable area until Auditing inspection as QAP reasonable appropriate attributes assurance is QC Inspection attained for that area 16

Enforcement Enforcement Enhancements Significance Enforcement o Risk-informed Determination 01.

Significance of o Builds on well-established Risk Insights SSC Non-approaches Conformance A B Deterministic o Leverages general reactor safety Risk Insights Criteria criteria vs facility-specific Finding/

+

quantitative risk assessment Violation C Severity 02.

o Significance determination effort Significance of Traditional Level Enforcement commensurate with risk QAP Breakdown o Appropriate level of detail to ensure clarity and consistency Significance of QAP Breakdown o Quantitative SDP maybe used for risk profiles and system complexity approaching LLWRs SSC = Structure, System & Component QAP = Quality Assurance Program 17

Enforcement Performance Assessment Enhancements Whats new under ARCOPs Two-Tiered Approach performance assessment? Risk Insights

01. Licensee Assessment Strategic Areas Based on severity level Informs supplemental Construction Quality of findings/violations & reactive inspections Security Programs (similar to cROP)

Operational Readiness Cornerstones 02. QAP Assessment Quality of Suppliers Enables timely NRC and Activities Assessment of QAP in SignificanceInforms of QAP changes Breakdown to baseline inspection plan licensee response to Construction, Manufacturing, each construction area performance deficiencies and Procurement Security Programs SSC = Structure, System & Component Operational Programs QAP = Quality Assurance Program18

ARCOP Development Timeline ARCOP guidance issued Address policy issues Inspection organization est.

ARCOP guidance begins Training 2022 Update IMC 2550 for adv. RTRs 2024 2026 SHINE OL Construction begins (NLWR Develop vision Hermes/ACU construction 2025 #1 & LWR-SMR #1)

Develop information SECY 2023 begins Int./Ext. Communication ARCOP guidance (con't)

Possible LWA issued 19

Advance Contracting Requirement Under Section 302(b) of the Nuclear Waste Policy Act Michael Kido Office of the General Counsel, U.S. Department of Energy March 2023

Nuclear Waste Policy Act of 1982 (as amended) (NWPA)

- Established the Federal responsibility for the disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW).

- Assigned to DOE the responsibility of developing capabilities for disposal and, if necessary, consolidated interim storage (referred to in the NWPA as monitored retrievable storage).

21

The Standard Contract - Background

- The Standard Contract for Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste (10 CFR Part 961) establishes the contractual terms and conditions under which DOE will make nuclear waste disposal services available to owners and generators of SNF and HLW (mostly nuclear utilities).

- The Standard Contract specifies the terms under which DOE will accept title to, transport and dispose of SNF and HLW from contract holders. It also provides for the payment of fees sufficient to offset DOEs expenditures.

22

Section 302(b) - Background

- Section 302(b) of the NWPA lays out the advance contracting requirement for NRC license applicants.

- NRC cannot issue or renew a license to any person to use a utilization or production facility under the authority of section 103 [Commercial Licenses] or 104

[Medical Therapy and Research and Development] of the Atomic Energy Act of 1954 unless such person has entered into a contract with DOE or DOE affirms in writing that such person is actively and in good faith negotiating with DOE for a contract.

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DOE Office of Standard Contract Management

-. DOEs Office of Standard Contract Management manages these Standard Contracts and the Nuclear Waste Fund for DOE. The Office is housed within the Office of the General Counsel and continues DOEs core functions established by the NWPA pertaining to the Nuclear Waste Fund and the management of the Standard Contract.

- The Standard Contract and the Amendment to the Standard Contract for New Reactors are available at the following link under Applicable Documents on this offices website - https://www.energy.gov/gc/office-standard-contract-management.

24

Questions?

25

Contact Information Michael Kido (DOE-OGC)

- Michael.Kido@hq.doe.gov Cyrus Nezhad (DOE-OGC)

- Cyrus.Nezhad@hq.doe.gov Connie Barton (Director, DOE Office of Standard Contract Management)

- Connie.Barton@hq.doe.gov 26

Advance Contracting Requirement Under Section 302(b) of the Nuclear Waste Policy Act -

NRC Guidance Joseph Sebrosky NRR/DANU/UARP Advanced Reactor Stakeholder Meeting March 2, 2023

Nuclear Waste Policy Act of 1982, as amended Section 302. Nuclear Waste Fund (b) ADVANCE CONTRACTING REQUIREMENT-(1)(A) The Commission shall not issue or renew a license to any person to use a utilization or production facility under the authority of section 103 or 104 of the Atomic Energy Act of 1954 (42 USC 2133, 2134) unless -

(i) such person has entered into a contract with the Secretary under this section; or (ii) the Secretary affirms in writing that such person is actively and in good faith negotiating with the Secretary for a contract under this section.

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Nuclear Waste Policy Act of 1982, as amended (continued)

(b) ADVANCE CONTRACTING REQUIREMENT [continued]-

(1)(B) The Commission, as it deems necessary or appropriate, may require as a precondition to the issuance or renewal of a license under section 103 or 104 of the Atomic Energy Act of 1954 (42 USC 2133, 2134) that the applicant for such license shall have entered into an agreement with the Secretary for the disposal of high-level radioactive waste and spent nuclear fuel that may result from the use of such license.

Source: NUREG-0980, Vol. 1, No. 10. ML13274A489.

29

Generic Letter No. 83 The Nuclear Waste Policy Act of 1982

  • Addressed to ALL POWER AND NON-POWER REACTOR LICENSEES, APPLICANTS FOR AN OPERATING LICENSE AND HOLDERS OF CONSTRUCTION PERMITS 30

NUREG-1537, Part 1, Rev. 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content

  • Published February 1996.
  • Section 1.7, Compliance With the Nuclear Waste Policy Act of 1982 The applicant should briefly discuss how it meets the requirements of Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 for disposal of high-level radioactive wastes and spent nuclear fuel. This discussion should include the contract arranged with DOE for return of the material. A copy of the cover letter for the contract between the applicant and DOE should be included in an appendix to the [safety analysis report].

31

Combined License Example

  • In Section 1.5.2 of each safety evaluation report on an AP1000 combined license (COL) application, the staff evaluates compliance with Section 302(b) of the Nuclear Waste Policy Act

- Example SER found at:

https://www.nrc.gov/docs/ML1227/ML12271A045.pdf

https://www.nrc.gov/docs/ML1527/ML15271A126.pdf) 32

Questions?

33

Micro-Reactor Licensing and Deployment Topics Advanced Reactor Stakeholders Meeting March 2, 2023 William Kennedy Amy Cubbage Advanced Reactor Policy Branch U.S. Nuclear Regulatory Commission

Introduction

  • Goals of this presentation
  • NRC draft white paper on micro-reactor licensing strategies
  • Licensing and deployment topics for factory-fabricated transportable micro-reactors
  • Discussion items 35

Goals of this Presentation

  • Inform stakeholders of the micro-reactor licensing and deployment topics currently being considered by the NRC staff for factory fabricated transportable micro-reactors
  • Hear feedback from stakeholders, including other topics for consideration and thoughts on prioritization 36

SECY-20-0093 Summary

  • SECY-20-00931 laid out several issues related to micro-reactor licensing and deployment, including information on the current regulations, applicability to micro-reactors, stakeholder perspectives, and NRC staff considerations
  • Some issues are being addressed in ongoing rulemakings and guidance development, and some are topics for consideration for factory-fabricated transportable micro-reactors as described later in this presentation 1 SECY-20-0093: Policy and Licensing Considerations Related to Micro-Reactors 37 (https://www.nrc.gov/docs/ML2025/ML20254A363.html)

SECY-20-0093 Summary

  • Security Requirements
  • Staffing, Training, and Qualification Requirements
  • Autonomous and Remote Operations
  • Regulatory Oversight
  • Aircraft Impact Assessment
  • Annual Fee Structure
  • Manufacturing Licenses and Transportation
  • Population-Related Siting Considerations
  • Environmental Considerations 38

Micro-reactor Licensing Strategies

- Enhanced standardization of the design and operational programs

- Manufacturing license may provide flexibility for design and fabrication in a factory and reduce site-specific inspections and verifications

- Use of bounding values for external hazards and site characteristics could reduce NRC staff review effort

- Generic Environmental Impact Statement for Advanced Nuclear Reactors (ANR GEIS) rulemaking 39

Licensing and Deployment Topics - Factory-Fabricated and Transportable Micro-Reactors

  • The NRC staff is continuing to develop topics related to licensing and deployment of factory-fabricated transportable micro-reactors to identify policy issues and options to address them
  • Loading fuel at a manufacturing facility Developers may propose loading fuel into reactors at the manufacturing facility either during or after the manufacturing process.
  • Qualifications for personnel handling fuel at a manufacturing facility Loading fuel at a manufacturing facility would also require appropriately-qualified personnel to handle the fuel.
  • Timelines for ITAAC closure, hearings, and 52.103(g) findings The process for beginning operation under combined licenses includes several steps with extended timeframes, such as ITAAC closure, the associated 52.103(g) finding, and the ITAAC hearing process (including the AEA 189a.(1)(B) requirement to provide notice of an opportunity for hearing at least 180 days before scheduled fuel load).

40

Licensing and Deployment Topics - Factory-Fabricated and Transportable Micro-Reactors

  • Licensing replacement of reactor modules Deployment scenarios may involve delivering fueled micro-reactor modules to the power plant site and replacing the modules with some periodicity.
  • Low Power Physics Testing at a Manufacturing Facility Developers may seek to load fuel and conduct low power physics testing at the manufacturing facility.
  • Transportation of fueled reactor modules Reactor modules that are loaded with fresh, irradiated, or spent fuel might be transported between the manufacturing facility, operating power plant site, and a facility for refurbishing or decommissioning reactor modules.

41

Licensing and Deployment Topics - Factory-Fabricated and Transportable Micro-Reactors

  • Remote and autonomous operations Micro-reactor developers might include capabilities for remote or autonomous operation and monitoring, including cybersecurity features, and propose not having on-site reactor operators.
  • Irradiated fuel and spent fuel The definition of spent fuel (10 CFR Parts 71 and 72) includes criteria that fuel has been withdrawn from a nuclear reactor following irradiation and has undergone at least one year's decay since being used as a source of energy in a power reactor. Depending on how long it has been since the final reactor shutdown of a micro-reactor, different regulations may apply to the storage and transport of the reactor fuel or the fueled micro-reactor module.
  • Decommissioning process/funding assurance Decommissioning transportable micro-reactors may involve independent regulated decommissioning of power plant sites as well as the reactor modules upon removal. Facility licensing and decommissioning licensing requirements may apply to developers who seek to use a centralized facility to decommission reactor modules away from power plant sites.

42

Additional Topics for Longer-Term Consideration

  • Mobile micro-reactors The NRC staff is aware that deployment of mobile micro-reactors is of interest to some developers.
  • Maritime or space applications The NRC staff is aware that maritime and space applications of micro-reactors may be of interest to developers.

43

Next Steps

  • Stakeholder engagement
  • Identify policy issues
  • Consider options to address the issues

- Guidance development

- Rulemaking

  • Draft White Paper to further stakeholder input
  • Engage Commission as appropriate 44

Discussion Items

  • Are there scenarios of interest that are not captured in this presentation?
  • What do stakeholders see as the highest priority topics to address?
  • Which regulatory topics pose the greatest risks to micro-reactor deployment?
  • Other feedback or questions 45

Advanced Reactor Stakeholder Public Meeting Lunch Break Meeting will resume at 1:00 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 417 405 578#

Transportation and Storage for Advanced Reactor Fuel and Transportable Microreactors Advanced Reactor Stakeholder Meeting March 2, 2023 Bernard White Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission

Introduction

  • NRC is ready to review transport packages and spent fuel storage applications

- Transportation package certification (10 CFR Part 71)

- Spent fuel storage installations (10 CFR Part 72)

  • NRC regulatory framework in 10 CFR Part 71 allows for the review of for advanced reactor fuel and transportable microreactors
  • NRC approved transportation packages and storage systems for TRISO and metallic fuels.

48

The Fuel Cycle Disposal 49

DFM Resources Thorough and timely reviews of advanced reactor package applications is a high priority for the NRC, and our reviews will ensure that new technologies may be used safely Early and frequent communication is key 50

Preparation

  • Training for NRC staff provides insights on significant safety features of specific designs and technologies
  • Technical reports addressing potential challenges assist staff in risk informing their reviews

- Review of Operating Experience for Transportation of Fresh (Unirradiated) Advanced Reactor Fuel Types (ML20184A151)

- Potential Challenges With Transportation Of Fresh (Unirradiated)

Advanced Reactor Fuel Types (ML20209A541)

  • Meetings with advanced reactor vendors provide staff with knowledge on specific designs and technologies 51

Preparation

  • NRC welcomes pre-application engagements to support an efficient review of new applications and amendments (LIC-FM-1, Overview & Expectations of the Certification and Licensing Process)
  • Early engagement helps NRC to understand future needs and inform its budget
  • NEI Letter dated December 15, 2020
  • Preapplication engagement ensures applicants and regulator have shared understanding of o the applicable requirements o review approach and o whether data gaps exist (e.g., testing) that need to be addressed.

52

Conclusion

  • NRC is proactively expanding our knowledge of advanced reactors and their fuels
  • Early engagement supports:

- a common understanding of the regulatory issues associated with advanced reactor fuel designs and technology

- Timely and efficient reviews

- NMSS and partners have sufficient resources

  • NRC review and oversight ensure safe use of transportation packages in the public domain 53

Bernard White, Yoira Diaz-Sanabria, Chief Sr. Project Manager Storage and Transportation Licensing Branch Bernard.White@nrc.gov Yoira.Diaz-Sanabria@nrc.gov 301-415-6577 301-415-8064 54

Guidance regarding Non-Power Liquid Fueled Molten Salt Reactor License Applications Advanced Reactor Stakeholder Meeting March 2, 2023 William B. Kennedy Senior Project Manager Advanced Reactor Policy Branch Division of Advanced Reactors and Non-Power Production and Utilization Facilities U.S. Nuclear Regulatory Commission

Contents

  • Background
  • Overview of the Oak Ridge National Laboratory (ORNL) report
  • NRC staff endorsement of Appendix A
  • Appendix B of the ORNL report
  • Future plans
  • Information resources 56

Background

  • Under contract with NRC, ORNL developed a report titled, Proposed Guidance for Preparing and Reviewing a Molten Salt Non-Power Reactor Application (ORNL/TM-2020/1478)

Overview of the ORNL Report

  • An information resource for stakeholders interested in licensing of non-power MSRs
  • Based on NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors
  • Focuses on the technical information needed to apply NUREG-1537 to the review of a non-power liquid fueled MSR license application 58

Overview of the ORNL Report

  • Main body describes the work to prepare the report
  • Appendix A, Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power MSRs: Format and Content
  • Appendix B, Part 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power MSRs: Standard Review Plan 59

Overview of the ORNL Report

  • Covers various topics, including:

- The facility

- Site characteristics

- Design of structures, systems, and components

- Molten salt reactor description

- Molten salt reactor cooling systems

- Engineered safety features

- Instrumentation and control systems

- Electrical power systems 60

Overview of the ORNL Report

  • Covers various topics, including:

- Auxiliary systems

- Experimental facilities and utilization

- Radiation protection program and waste management

- Conduct of operations

- Accident analyses

- Technical specifications

- Other license considerations 61

Overview of the ORNL Report

  • Refers to existing guidance in NUREG-1537 and interim staff guidance augmenting NUREG-1537 for other topics:

- Financial qualifications

- Decommissioning

- Environmental review 62

NRC Staff Endorsement of Appendix A

  • By letter dated November 18, 2020, the NRC staff endorsed Appendix A of the ORNL report as guidance, subject to certain clarifications, for preparing license applications for non-power liquid fueled MSRs under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Section 50.21(c).

(https://www.nrc.gov/docs/ML2025/ML20251A008.pdf)

  • Helps applicants provide the information required by 10 CFR 50.34, Contents of applications; technical information, and other regulations 63

Appendix B of the ORNL Report

  • Appendix B provides a standard review plan tailored to liquid fueled molten salt reactor technology, including:

- Areas of review

- Acceptance criteria

- Review procedures

- Evaluation findings

- Technical rationale 64

Future Plans

  • The NRC staff is considering whether to endorse Appendix B as guidance in the near term
  • In the longer term, the NRC staff plans to incorporate the ORNL report, as appropriate, in a new volume of NUREG-1537 covering non-power liquid fueled MSRs 65

Information Resources

  • Endorsement of Appendix A to Oak Ridge National Laboratory Report Titled, Proposed Guidance for Preparing and Reviewing a Molten Salt Non-Power Reactor Application, as Guidance for Preparing Applications for the Licensing of Non-Power Liquid Fueled Molten Salt Reactors (ADAMS Accession No. ML20251A008) 66

Information Resources

  • Proposed Guidance for Preparing and Reviewing a Molten Salt Non-Power Reactor Application (ORNL/TM-2020/1478) (ADAMS Accession No. ML20219A771)
  • NUREG-1537, Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Contents (ADAMS Accession No. ML042430055)
  • NUREG-1537, Part 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review Plan and Acceptance Criteria (ADAMS Accession No. ML042430048) 67

Questions?

Contact me by e-mail at William.Kennedy@nrc.gov or by telephone at (301) 415-2313 68

Pre-Application Engagement on Materials Qualification Issues for Advanced Reactor Licensing Meg Audrain Office of Nuclear Reactor Regulation March 2, 2023 Advanced Reactor Stakeholder Meeting

Agenda

  • Why have pre-application engagements?
  • Code Requirements
  • Environmental Testing
  • Design Envelope
  • Non-Code Qualified Materials 70

Why Have Early Engagement?

  • Encouraged for all materials used in safety related and risk-significant applications
  • Important to ensure NRC staff and applicants have a common understanding on data requirements for these materials
  • More efficient for applicants and NRC staff to do this in pre-application space to ensure timely application reviews.

- Significant lead time for materials testing could lead to delays if not addressed early 71

ASME Code Requirements

  • NRC staff anticipates most applicants will qualify materials and designs to ASME Section III, Division 5, because many of the proposed designs operate at temperatures or in environments where existing Codes endorsed by the NRC or incorporated by reference do not apply.
  • Applicants should demonstrate how their design complies with Div 5, as conditioned in RG 1.87
  • Applicants should justify deviations from Div 5 and demonstrate why the proposed deviations are acceptable. The use of alternative codes of construction should include a delta analysis.

72

Environmental Testing

  • Div 5 rules do not cover deterioration that may occur in service as a result of radiation effects, corrosion, erosion, thermal embrittlement, or instability of the material but states that these effects shall be taken into account for design or service life
  • NRCs forthcoming Materials in Advanced Reactors ISG provides guidance for staff reviews in this area. Applicants should consider this information as they develop qualification, monitoring and surveillance programs
  • Environmental testing data is potentially time consuming to gather and results could impact design or component lifetimes 73

Environmental Testing

  • Used to develop corrosion or degradation rates for specific reactor environments
  • Needed to understand environmental effects and their impacts on mechanical and thermal behavior
  • Needed to set appropriate limits on coolant purity

- Not explicitly addressed like it is for LWRs

- No coolant purity standards exist for non-LWR environments

  • Needed to determine if transient could potentially be end of life event 74

Data Supports Design Envelope

  • Should show that any data used, historic or planned, is directly applicable to plant design and environment
  • Data should support design for operating and accident conditions
  • Confirm that any standards referenced in Div 5 were used or provide a delta analysis for standards that were used (e.g., QA programs) 75

Use of non-Code Qualified Materials

  • For use of non-Code Qualified materials, the NRC will review material qualification data

- ensure material and mechanical properties support intended functions

- environmental testing still needed

  • Applicants should demonstrate that graphite will be qualified as per Div 5. In addition, any deviations from Code should be addressed 76

Conclusions

  • Early engagement is important to support timely application reviews
  • NRC wants to ensure a common understanding on data qualification and any potential testing requirements during pre-application
  • Beneficial to both NRC staff and applicants 77

Questions?

Interim Staff Guidance on Materials Compatibility in Advanced Reactor Environments Meg Audrain Office of Nuclear Reactor Regulation March 2, 2023 Advanced Reactor Stakeholder Meeting

Agenda

  • Public Comment Period
  • NRC Stakeholders
  • Applicability and Purpose of ISG
  • Regulatory Framework
  • Qualification and Performance Monitoring
  • Technical Content
  • Conclusions and Questions 80

Public Comment Period

  • Draft ISG, Material Compatibility for Non-Light Water Reactors, DANU-ISG-2023-01 (ML22203A175)

- FRN will be published in early March 2023

  • 60-day public comment period: early March - early May 2023
  • Submit comments to be considered by staff. Only written comments will be formally addressed in the final ISG.

- www.regulations.gov; Docket ID NRC-2022-0215 81

Why Develop the ISG?

  • Staff expects that most applicants will demonstrate their materials meet ASME Section III, Division 5 (Div 5), High Temperature Reactors
  • Div 5 rules do not cover environmental combability; however, it states that these effects shall be taken into account for design or service life of structures, systems and components (SSCs)
  • Currently no staff guidance on how to review materials qualification, performance monitoring methods, and surveillance for non-LWRs
  • Staff guidance will ensure consistency and clarity for reviewing applications

- Identify information related to materials qualification that the NRC staff should consider in their reviews

- Guide the staff in identifying where monitoring and surveillance programs may be appropriate 82

Applicability

  • Applicable to NRC staff reviews of non-LWR designs that propose to use materials allowed under Div 5

- Power and non-power reactors

- Part 50 - construction permit and operating license

- Part 52 - design certification, combined license, standard design approval, or manufacturing license 83

Non-LWR environment

  • Non-LWR environments may have unique material corrosion, degradation mechanisms, and irradiation effects
  • Studies have identified the gaps in knowledge that exist for some of these coolant types and the impact on the materials being considered in the construction and operation of these non-LWR nuclear power plants
  • Because of the state of knowledge and long test times, there is a strong emphasis on using mitigation strategies, performance monitoring, and surveillance programs to ensure SSCs continue to satisfy the design criteria 84

Current Regulatory Framework

  • For non-LWRs, Regulatory Guide (RG) 1.232, Guidance for Developing Principal Design Criteria for Non-Light Water Reactors, issued March 2018, provides proposed guidance for the development of principal design criteria for non-LWR reactors
  • Several design criteria relate to materials qualification for structural materials and state the importance of environmental compatibility, inspection, materials surveillance and functional testing 85

Qualification and Performance Monitoring -

Terminology

  • Materials qualification

- Testing conducted in an environment simulating the anticipated operating environment for the reactor, including chemical environment, temperatures, and irradiation

  • Performance monitoring

- Inspections or examinations to confirm adequate performance and to identify unacceptable degradation

- May also include aging management programs or post-service evaluations

  • Surveillance programs

- Examination of test coupons and components removed from the reactor over the licensed operating period 86

Qualification and Performance Monitoring

  • An SSCs performance will be demonstrated through a combination of materials qualification programs, performance monitoring, and surveillance programs, which collectively provide assurance that a component will meet the design requirements over its intended design life in the applicable environment
  • The scope of materials qualification and monitoring programs should include safety-related component materials, safety-significant component material, and as needed, non-safety related component materials whose failure could impact critical design functions
  • Testing should be conducted to determine if materials properties and allowable stresses meet applicable codes and standards or other design requirements 87

Qualification and Performance Monitoring

  • Availability of data on performance in a specific operating environment will inform the review to ensure an SSC will maintain its intended function

- Little data - could require robust performance monitoring and surveillance programs

- Large amount of data or significant design margin - may require less rigorous performance monitoring and surveillance programs

  • Performance monitoring and surveillance programs could be needed for SSCs that are not planned to undergo periodic inspections and/or functional testing 88

Technical Content of ISG

  • The ISG separates degradation issues into generically applicable issues and technology specific issues
  • Three technology specific appendices

- Molten salt reactors, liquid metal reactors, and HTGRs

  • Represents current state of knowledge - as additional operating experience and laboratory testing become available, treatment of issues may change, and new issues may be identified.

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General Degradation Mechanisms

  • Corrosion
  • Creep and creep Fatigue
  • Environmentally assisted cracking
  • Flow induced degradation (abrasion, erosion, cavitation)
  • Gaskets and Seal chemical compatibility
  • Irradiation effects
  • Stress relaxation cracking
  • Wear/fretting 90

General Materials Issues

  • Advanced manufacturing technologies
  • Lubricants
  • Ceramic insulation
  • Weld design and fabrication
  • SiC/SiC composites
  • SA-508/533 Bainitic Steel for RPVs 91

Molten Salt Reactor Appendix

  • Graphite compatibility
  • Materials considerations (degradation, cracking, corrosion)
  • Salt composition

Liquid Metal Reactor Appendix Sodium-cooled fast reactors Lead-cooled fast reactors

  • Caustic stress-corrosion cracking
  • High temperature corrosion
  • Exothermic reactivity with water
  • Effect of flow velocity
  • Liquid metal embrittlement corrosion
  • Nonmetallic materials
  • Combining ferritic steels and
  • Oxygen control austenitic steels (galvanic corrosion)
  • Liquid metal embrittlement 93

High Temperature Gas Cooled Reactor Appendix

  • Creep-rupture strength
  • Emissivity
  • Graphite
  • Graphite dust
  • Helium impurities
  • Metallic materials qualification considerations 94

Conclusions

  • NRC staff developed an ISG to guide staff on reviewing applications using materials allowed under Div 5
  • ISG has been issued for public comment

- Comment period - March to May 2023

  • NRC staff encourages stakeholders to provide feedback on contents of ISG through this process 95

Questions?

Advanced Reactor Stakeholder Public Meeting Break Meeting will resume at 2:45 pm EST Microsoft Teams Meeting Bridgeline: 301-576-2978 Conference ID: 417 405 578#

Periodic Advanced Reactor Stakeholder Meeting: Status of Draft Regulatory Guide 1410 and 1307, Including Responses to NEI Comments Dr. John Stamatakos Institute Scientist at Southwest Research Institute March 2, 2023

Overview

  • Changes since publication of the Pre-decisional guides

- Current versions address both Framework A and Framework B, consistent with the most recent version of 10 CFR Part 53

- Three options apply to both frameworks

  • Discuss NEI comments and responses (four main groups)

- Comments that were addressed/incorporated in the current drafts

- Comments used for planning Appendix B and to revised RIL 2102-04/NUREG for Option 3.

- ASCE 7 related

- Part 50/52 related

  • Future plans and summary 99

Changes Since Publication of Both The Pre-decisional Guides

  • Added discussions in sections A and B for both Framework A and Framework B, consistent with the most recent version of 10 CFR Part 53 (prior draft was only for Framework A)
  • Modified all three options to address both frameworks
  • Incorporated many review comments and suggestions, including those from NEI.

100

Table 1 and Table 2 (next slide) are from Draft RG 1410 but are applicable to both RGs.

101

Table 1 and Table 2 (next slide) are from Draft RG 1410 but are applicable to both RGs.

102

Pre-decisional Draft RG 1307

- Technical considerations:

  • Use the same technical approach as described in Pre-decisional Draft RG 1407 (3 options)
  • Focus on addressing SI specific criteria for each of the 3 options

Revised flowchart from Draft RG 1307 104

Color Coding to Categorize Responses in NEI Table We organized the NEI comments in a table and then categorized them as follows:

Description Comment #s Comment #s (Part 1) (Part 2)

To be incorporated in the next 1, 3, 5, 7, 9, 1, 2, 3, 5, revision of the RGs 12, 13 7, 12, 14 To be addressed in next revision of 4, 6, 14, 18, 19 9, 10, 11, 13 RIL/ NUREG We feel outside the scope of the RGs 2, 8, 10, 11, 4, 6, 8 15, 16, 17 105

Example: Comments that Were Incorporated in the Revised RGs 106

Appendix B

  • Working with NRC Staff, the SwRI team will develop examples of how to implement Option 3 (with specific ties to Framework A and Framework B as necessary).
  • Option 3 provides flexibility to an applicant for seismic design considering unique aspects of its plant design, site, and other considerations. We will focus on design and analysis strategies that an applicant can follow to demonstrate compliance with the safety and risk requirements in 10 CFR Part 53 using Option 3.
  • As necessary, we will demonstrate key steps in our example strategies with performance and risk analyses similar to the ones already provided in Appendix A.
  • More details will be developed in a revision to RIL 2021-04 that is expected to be in the form of a NUREG/CR to support the two RGs.

107

Example: Comments To be Addressed in the Next Revision of RIL 2021-04/ NUREG 108

Example: Comments We feel Are Outside the Scope of the RGs 109

10 CFR Part 50 and 52

  • There are no longer any references to these regulations in the two Draft RGs.
  • NRC staff will evaluate the potential to develop additional guidance on how the RIPB approaches using ASCE 43-19 and ASCE 4-16 can be adopted under these regulations

110

Example: Comments We feel Are Outside the Scope of the RGs 111

ASCE 7

  • ASCE 43, ASCE 4, and associated design codes reflect current Nuclear Industry design and construction practices and produce acceptable design with sufficient margin (actual performance is a function of both design and construction) as demonstrated by recent SPRAs.
  • NRC has evaluated ASCE 43 and ASCE 4 in detail and has developed regulatory positions with exceptions, additions, and clarifications.
  • ASCE 7 takes a different approach to safety and performance, and NRC staff (and industry to our knowledge) have not yet evaluated how to align this approach with the current NRC regulatory approach for power production commercial nuclear plants.
  • The Draft RGs provides an acceptable way to meet the regulations. Therefore, the following statement in RG: any code other than ASCE 43-19 and ASCE 4-16 for seismic design of SSCs with appropriate justification.
  • We propose a technical meeting to discuss what information is needed to evaluate ASCE 7.

112

Summary

  • Revised Draft RG 1410 and Draft RG 1307 have been updated to include both Framework A and Framework B, and all three options are now available for both frameworks.
  • The Draft RGs will be revised to address comments received for the Trial DG including adding an Appendix B to illustrate Option 3.
  • We have addressed NEI comments and plan to address and incorporate all public comments that fall within the scope of the two Draft RGs.
  • We anticipate Appendix B draft in 3 months (end of May 2023).

113

Future Meeting Planning

  • The next periodic stakeholder meeting will be scheduled for April or May 2023.
  • If you have suggested topics, please reach out to Steve Lynch at Steven.Lynch@nrc.gov 114

How Did We Do?

  • Click link to NRC public meeting information:

https://www.nrc.gov/pmns/mtg?do=details&Code=20230075

  • Then, click link to NRC public feedback form:

115

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